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UCRL 89830 PREPRINT

f + b S O 7 - - 9

Nuclear Waste Package Design f o r the Vadose Zone i n T u f f

UCRL--898 30

DE84 007791

ANS/ASME Waste Management '84 Tiicson, Arizona

March 11 - 15, 1984

February 1984 I

This is a preprint of a paper intended for publication in a journal or proceedings. Since 7 changes may be made before publication, this preprint is made available with the understanding that it will not be cited or reproduced without the permission of the author.

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Nuclear Waste Package Design f o r tne Vadose Zone in Tuff* -3

W. C. O'Neal, L. B. Ballou, D. W . Gregg, E. W . Russell

Lawrence Livermore National Laboratory

Livermore, CA 94550

ABSTRACT :

Tnis report presents an overview of the selection and analysis of conceptual waste package designs tna t will be used by tne Nevada Nuclear Waste Storage Investigations (NNWSI) project f o r disposal of n i g n level nuclear

waste ( H L W ) a t the proposed Yucca Mountain, Nevada Si te .

The design requirements t ha t the waste packages are required t o meet are

Concept drawings f o r the reference designs and one alternative l is ted.

pacKage design are shown. Four metal alloys; 304L SS, 321 SS, 316L SS and

Incoloy 825 nave been selected f o r candidate canister/overpack materials, and

1020 carbon s teel has been selected as tne reference metal f o r the borehole

l iners .

A summary of the resu l t s of technical and economic analysis supporting tne

selection of the conceptual waste pacKage designs i s included. containment and release rates are not discussed in t n i s paper.

Post-closure

*Wow performed under tne auspices of the U.S. Department of Energy by tne * Lawrence Livermore National Laboratory under contract number W-7405-Eng-48.

In troduct i on

Under the d i r fion of the Offi of Ci v i 1 i an Radi oa t i ve Waste Management, the Department of Energy's Nevada Nuclear Waste Storage Investigations (NNWSI) project i s evaluating a candidate s i t e a t Yucca

Mountain, Nevada, f o r permanent disposal of h i g n level nuclear waste. '

Lawrence Livermore National Laboratory ( L L N L ) , a participant i n tne NNWSI

project, i s developing waste package designs t o meet NRC requirements. Included are designs f o r the current reference waste form configurations of:

1 ) Spent Fuel (SF) , which consists of both consolidated and unconsolidated

spent fuel rods from Pressurized Water Reactor (PWK) and Boiling Water Reactor

(BWR) assemlies , 2 ) Commercial High Level Waste ( C H L W ) , a s a borosilicate glass containing commercial spent fuel reprocessing wastes, and 3) Defense

Hign Level Waste ( D H L W ) immobilized i n borosi1:icate glass.

a l ternat ive designs have been developed f o r each waste form f o r bo tn vertical

and norizontal emplacement configurations. All designs are f o r emplacement in

a tuff repository located above the water tab le i n the vadose zone.

Reference and

Conceptual designs and analyses fo r waste packages in tuff below tne water

tab le were developed f o r Office of Nuclear Waste Isolation (ONWI) by

Westinghouse Electric Corporation in 1981-82 (Schornhorst 1983).

horizon was changed by NNWS1:to tne vadose zone i n l a t e 1982 (Vietn 1982, Dudley 1982).

Tne target

LLNL has made changes and additions t o the conceptual designs,

and analyses were performed t o dete'rmine conformdnce of the selected design ensemble t o NRC design requirements in the currently understood repositqry

environment.' Figure 1 shows the current reference conceptual designs. Tne

selected designs (Gregg 1983) include reference and al ternat ive designs which

vary in complexity, performance and cost.

From tnis ensemble, one s e t of designs (for Spent

will be chosen vJhich should meet NRC/EPA requirements

accurate repository data and long term corrosion and developed (McCrignt 1983, Oversby 983).

Fuel, CHLW, and DHLW)

wnen analyzed witn

each r a t e data now be

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I r 7 1 cm dia p-71 cm dia?

16.5

4

DHLW package empl aced in verti cal borehol e.

CHtW package empl aced i n vertical borehol e.

Spent fuel waste package emplaced i n vertical borehol e.

Fig. 1. Reference Waste Package Design. comnercial high-level waste, the reference design consists of 304L pour canisters w i t h no overpacks and no packing. the reference design canister is one cm thjck 304L w i t h no overpack and no packing. The reference emplacement configuration is vertical emplacement beneath d r i f t f loors .

For defense high-level waste and

For spent fuel ,

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The conceptual designs considered to date do not include considerations for TRU waste packages. information on the characteristics of TRU waste forms is available.

These will be developed in the future when more

Waste Package Environment

Tne candidate repository horizon is located in dense-welded tuff, 350 to 400 meters below the surface, and approximately 100 meters above the water table. Our current understanding of the repository environment for expected conditions is: the hydrostatic and lithostatic forces on the waste package will De zero; there will potentially be some load (nominally 0.1 MPa) on canisters due to caving in o.f borehole walls, or one-meter s zed rock blocks falling in; there will be one atmosphere of air; and about 8 mm per year of water dripping on or seeping around the waste packages (Sass 1982). decay radiation heats the borehole wall to temperature above 95OC, the packages will experience a steam-air mixture. 95'C, if water has resaturated the rocK, water may drip onto pour canisters and evaporate for tens of years (spent fuel canisters stay above 95OC for several hundred years in these designs) (Hockman 1983). alter the chemistry of the air and water around the canisters (McCrignt 1983).

When

When the borehole walls cool to

Gamma radiation may

Pre-ClOSUre Environment and Effects on Canisters

The pre-closure environment for DHLW and CHLW pour canisters begins with the glass pouring operation. For DHLW, the reference canister design is a single barrier, one-cm (3/8 in) thick 61 cm (24 in) diameter 304L container which is filled with a molten glass and waste mixture and emplaced with no overpack. Differential thermal contraction and deformation between the 304L SS wall and hot glass during cooling produces a complicated stress distribution in the canister wall (Baxter'1983, Slate 1981). Tne time/temperature hi story and accumulated residual stress pattern may cause sensitization (carbide precipitation) and create conditions leading to stress corrosion cracking susceptibility after emplacement (McCright 1983, Slate 1981).

canister, except for diameter which is 32 cm (12 in). stresses could be minimized in pour canisters by utilizing a thin-walled inner

The reference CHLW pour canister design is identical to tne DHL'UI Residual thermal

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. . . ,

. , . i:

liner with clearance. overpacking and could be utilized if tne magnitude of residual stresses in pour canister walls proves unacceptable.

Tnis would be consideraoly less expensive than I

The shipping and handling environment of the stressed DHLW and CHLW pour canisters probably will not cnange the physical or chemical state of the Canisters provided the canisters are properly fixed in tneir shipping casks and snipping procedures are followed. possible cracking due to fatigue or creep during transportation has not yet Deen undertaken. At the repository, the normal handling, transporting and emplacement environment should not overstress the pour canisters. Retrieval operations could produce additional stresses in pour canisters depending on the integrity of tne emplacement system at retrieval. stresses imposed during retrieval will be superimposed onto existing thermally induced stresses from glass pouring operations. canister is fabricated from 304L stainless steel with a wall thicKness of one cm. canister overpacks the normal emplacement and retrieval operations do not produce stresses above allowable limits.

Documented analysis and testing for

For pour canisters,

Tne reference spent fuel

End caps are 3.8 cm tnick. For these spent fuel canisters and for pour

Fire test and drop test environments for spent fuel canisters and pour canisters will impose nigh stresses. stresses (Ross 1979) and reports on drop tests of spent fuel canisters (Kurasch 1980, Slate 1981) indicate tnat reference designs for SF canisters and pour canister overpacKs will meet requirements for these tests. deformations occurring during drop tests of unlined pour canisters will De superimposed onto deformations which took place in pouring operations. Temperatures up to 8OO0C for 30 minutes will occur in fire tests. heating of pour canisters increases internal gas pressure but may actually

Calculations and data on fire test

Plastic

Fire test

lower the residual tensile stress in tne canister wall as it expands away from tne glass.

Design Objectives, Philosopny and Description

The objective is to develop and analyze waste package designs wnich incorporate qualified materials and which are fully compatiole with the

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repository design. dpplication by demonstrating conformance with requirements for safe narlaling, transportation, emplacement, retrieval, containment and release rate per NRC 10 CFR 60 and 10 CFR 71.

Tne designs and analyses are needed to support license '

Tne basic design philosophy for waste package conceptual designs is to meet NRC design criteria with flexibility in' technical performance and cost. This provides for the present uncertainties in repository environment and corrosion and release rate data. reference designs which are simple and economical, and alternative designs with greater tecnnical conservatism but added complexity and higher cost. designs are generally suitable for vertical or horizontal emplacement witn mi nor modifications .

To accomplish this, we nave selected

All

Tne simple designs specify direct burial of one-cm wall 304L stainless steel pour canisters and spent fuel canisters. Alternative designs specify more corrosion resistant alloys such as 316L SS, 321 SS and Incoloy 825, for containment barriers. Alternative designs are considered for internally lined pour canisters t o reduce residual wall stresses. Designs have been prepared wnicn include overpacks and packing materials (Gregg 1983). These alternative designs are more conservative and more costly than tne reference designs, but may De necessary to meet NRC criteria if long term corrosion and release rate testing produces rates significantly higher than currently availaDle snort term data indicate..

Design Requirements and Constraints

We are designing waste packages to meet a set of NNWSI requirements derived from NRC 10CFRGO and NRC 10CFR71. (Russell 1983).

Table 1 lists the requirements

TABLE 1. NNNSI Design Requirements Derived from NRC lOCFR6O.

bJaste packages shall be designed to:

1 . Contain the Naste for 300 to 1000 years.

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2.

3.

4.

5.

6.

7.

8.

9.

Maintain a release r a t e less than

inventory present a t the end of the containment period (300 years minimum).

Be retr ievable fo r 50 years a f t e r emplacement of tne f i r s t waste pacKage.

Meet nuclear c r i t i c a l i t y safety standards, i .e., n o t exceed an effect ive

mu1 t ip l ica t ion factor ( K e f f ) of 0.95.

Not exceed temperature l imi t s of the waste forms, which are 773 K (500°C)

for DHLW glass, 673 K (400°C) f o r CHLW glass, and 623 K (350°C) f o r spent

fuel cladding.

Not leak radioactive material in excess of applicable federal and s t a t e

standards a f t e r a d r o p t es t of two times waste package length onto an

unyielding surface, a t the minimum anticipated temperature.

Not leak radioactive material in excess of applicable federal and s t a t e

standards a f t e r sustaining a 1073 K (800"C), 30-minute f i r e test .

Not leak radioactive material in excess of applicable federal and s t a t e standards during or a f t e r transportation, handling, emplacement,

re t r ieva l , and expected seismic loads. Further, these loads must not compromise 1 ong-term performance.

Recain legible, externally labeled ident i f icat ion as long as

re t r ievabi l i ty i s required.

per year of the radionuclide

10. Meet federal regulatory requirements f o r transportation of high level

11. Meet requirements w i t h consideration f o r cost-effectiveness, including

nuclear waste.

d i rec t package costs and related repository system costs througn the

operat i onal period.

Nuclear Cr i t i ca l i t y Safety Analysis of ConceDtual Desisns

A c r i t i c a l i t y safety assessment was made for ieference DHLW canis ters

(Baxter 1983). The resu l t s show tna t the highest calculated c r i t i c a l i t y

coeff ic ient was k a = 0.147. This i s well below the maximum allowable of 0.95.

However, f o r DHLW and CHLW, the f i s s i l e material content of Keff t n e waste forms proposed t o date i s suf f ic ien t ly low t n a t a c r i t i c a l i t y

accident i s n o t of concern during the containment period (Schornhorst 1983, p

60).

The calculations nad n o t included flooding conditions which may incrase

s l ight ly .

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For various dry and flooded configurations of spent fuel canisters emplaced in the tuff repository, recent criticality calculations snow that

1983). The calculations indicate that keff will be less than 0.95 unless, as required in NRC lOCFRGO,.two unlikely, independent, and concurrent changes have occurred in the conditions essential to nuclear criticality safety. The Changes necessary to exceed keff 0.95 are:

is always below 0.95 for spent fuel depleted to ~1.4 w/o U235 (Weren Keff

The emplaced canister is loaded with spent fuel with equivalent (depleted) loading gt-eater than 1.4-1.6 w/o U235. (Undepleted fuel rods will be specially handled in the repository. Fuel assemblies are normally depleted to 1 .O-1.5 w/o.) and; , The canister is breached and filled witn water and; The spent fuel rods or space frame have disintegrated and; a. The spent rods have rearranged into an optimal configuration

(>1.6 w/o U235) or; Tne Zircaloy cladding has disintegrated and all the spent fuel

has fallen into a pile (>1.4 w/o U235).

b .

Thermal Modelina

Calculated waste package temperatures. for a1 1 of the reference and alternative designs are within tne temperature limits imposed to avoid waste-form degradation (Hockman 1984). These 1 imi ts are 350°C for spent fuel

. (cladding), 400°C for CHLW and 500°C for DHLW. consolidated spent fuel, the same diameter canister (65 cm) will accomodate 18 BWR assemblies (3.42 kW) and 7 PWR assemblies (3.56 kW). SF canister witn pawing, it is necessary to reduce the number of assemplies to 4 PWR (2.0 KW) to avoid exceeding the 350°C limit. containing 6 PWR assemblies of pre-consolidated (boxed) spent fuel has a peaK temperature of 335°C.

In the canister design for

For the alternative

The 68 cm canister

Results of. thermal analyses show that the reference package designs for CHLW and DHLW when emplaced in the reference repository geometry (vertical einp 1 acement ) are we1 1 bel ow waste form temperature ,1 imi t s . and DHLW canisters could be emplaced at nigher areal power density without

In fact, both CHLW

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exceeding temperature l imits.

detailed configurations and fuel loadings, the reference design spent fuel canisters can be emplaced i n the reference repository and not exceed waste

form temperature 1 imi ts.

The analysis also shows tha t , depending on

Structural Analyses of Canister Designs

The pre-closure environmental history of canisters and overpacts begins

with canister fabrication and continues through glass pouring (DHLW and CHLW

on ly ) , transportation, handling, welding, emplacement, and possible

re t r ieval .

w i t h o u t canister leakage.

normal , t e s t , and unexpected environmental conditions which will be imposed on canisters.

t e s t s (Ross 1979) have been conducted on these types of canisters by several

organiza t ions . survive such t e s t s .

For prototype canis ters , f i r e and d r o p t e s t s must be completed

Structural analysis is underway t o simulate the

Drop t e s t s (Ross 1979, Slate 1981, Kurasch 1980) and overheat

Based on these resul ts , NNWSI canis ter designs are expected t o

Fire t e s t computer simulations were performed f o r a spent fuel canister

using TAC02D (Burns 1982) t o calculate internal temperatures from an 8OO0C, 30 minute boundary condition. These temperatures were used t o calculate internal

pressures. The internal pressure was applied on a boundary condition t o model

Canister s t resses using NIKE2D (Hallquist 1979). Assuming increased pressure

from fuel-rod gas release, tne maximum s t r e s s i n the canister was 9,900 psi.

Tnis gives a safety factor of two on ultimate strengtn.

Structural analysis f o r l i f t i n g loads u s i n g NIKEZD gave a maximum s t r e s s

of 4,400 psi (30.3 MPa) ( i n the p in t le ) . A retrieval load of 80,000 l o

(356 k N ) can be applied on the pint le w i t h o u t exceeding tne 304L stainless

s teel yield strength of 30,000 psi (207 MPa) i n the p in t le or canister.

To simulate tne (unexpected) flooding condition i n the repository, a hydrostatic pressure was applied t o the outside of the canister.

strength was not exceeded u n t i l 230 psig (1.59 MPa) of water pressure was applied. Tnis pressure would resu l t from a water head of over 500 f t ( 164

m ) . incidents, and f o r significant loadings from loca rock fa i lures .

Tne yield

Canister strength i s tnerefore suff ic ient t o allow f o r unlikely f l o o d i n g

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Structural analysis of other canister designs and conditions is continuing. requirements over the expected range of pre-closure environments.

The results thus far indicate that all designs will meet strengtn

Design Description

Reference Designs

The reference conceptual designs represent the east complicated configurations (see Fig. 1). The DHLW 61 cm diameter pour canister and the 32 cm diameter CHLW pour canister are both nominally one cm thick, 304L stainless steel with identical pintles of the Savannah Rivdr type (Baxter 1983).

The reference spent fuel canister i s 65 cm diameter with 4.0 and 4.5 m lengths (plus pint1e)to accomodate various length spent fuel rods. canister contains 7 PWR or 18 BWR assemblies. For intact assemblies and preconsolidated boxed spent fuel, the diameter i s 68 cm and the lengths are 4.25 and 4.75 m (plus pintle). stainless steel with wall thickness of one cm. The pintles (16.5 cm) are identical to pour canister pintles.

The

The canisters are fabricated from 304L

The internal space-frame for spent fuel canisters provides mechanical stability and enhances heat transfer. For consolidated rods, the space-frame consists of 12-14 one cm thick, carbon steel radial fins in a cylindrical shell witn end plates. space-frame consists of a one cm thick "pigeon-hole" array with square receptacles and end plates. preconsolidated PWR assemblies. For BWR assemblies, it contains nine intact or 18 preconsolidated assemblies.

boxed assemblies is contingent upon the successful "two for one" volume reduction envisaged and demonstrated for preconsol idation (Anderson 1982).

For preconsolidated or intact assemblies, tne

The PWR canister contains three intact or six

The 68 cm diameter for preconsolidated

None of the reference designs utilizes an overpack container. All are intended for single package emplacement in a vertical borehole.

I

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Alternative Desians

Alternative metals (under 'long term test ing along w i t h 304L) are 316L SST,

321 SST and Incoloy 825 nickel-base alloy. These represent more

corrosion-resistant b u t more expensive metals. Results of long term test ing will be used t o determine the alloy selection and container thickness (Russell

1983, McCright 1983). A m i n i m u m thickness of one cm i s required for strength.

The a1 ternat i ve empl acement mode i s horizontal empl acement. A carbon

s tee l l iner will be used t o keep the borehole f r e e from sloughing rock t o

f a c i l i t a t e emplacement and retr ieval . Reference canister designs can be

emplaced horizontally. Figure 2 shows the al ternat ive spent fuel package

design w i t h packing emplaced vertically.

Residual thermal s t resses or sensit ization i n 304L pour canisters may

If so, three design alternatives have been developed: prove unacceptable.

s t r e s s corrosion res i s tan t a1 loys, internal 1 iners, and overpacks. When

tes t ing data and analysis resu l t s are obtained, i t will become more c lear as

t o which a l te rna t ive(s ) i s most appropriate.

Economi c Anal ys i s

Costs have been estimated for a l l reference and al ternat ive designs. The

most cost effective designs are those tha t maximize the canister s ize without exceeding waste f o r i temperature limits: The most cost ly a l ternat ive i s spent

fuel w i t h packing, ver t ical ly emplaced (see F ig . 2 ) .

package design related costs of reference designs and tne al ternat ive design

spent fuel package ( 4 PWR assemblies) w i t h packing. a re included. Costs fo r underground and above ground f a c i l i t i e s (buildings,

equipment, roads, drifts , boreholes, access shafts, e tc . are not included).

Table 2 gives waste

Most waste package costs

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Fipure 2

-Pintles

I

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' spent fuel - PWR - four assemblies with eight fins

Alternative design for consolidated spent fuel waste packages if packing (backfill) is needed to reduce leach rate.

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TABLE 2

High Level Waste Package Costs (1983 Dollars) 304L Canister, Vertical Emplacement

Reference 7 PWR

DHLW CHLW Spent Fuel

Power, kW 0.42 2.21 3.56 Canister $5200 3200 9500 Packing Repository Processing 3800* 3800" 8700* Consolidation --- --- 24000* Radiation P l u g 6400 5 200 7 400

Total cos t per Pkg $15400 $1 2200 $49600 No. Pkg per y r 5 00 6 60 288 Annual cos t $7.7M $8.1M $1 4M

* Schornhorst 1983

--- --e ---

Conclusions

A 1 t e rna t 4 PWR

ve

Spent Fuel with Packing

2.01 7900

14700 6600

14000 10700

$53900 504

$27M

T h e conceptual design ensemble provides a group of designs from which a

s e t can be chosen f o r eventual detai led des ign of waste packages. Within the

var ia t ions in materials and configurations available, detai led designs should

meet a l l requirements, barring unexpected results from long-term materials

t e s t ing or package environment characterization a c t i v i t i e s .

References

Anderson, R . T . , 1982, "Studies and Research Concerning BNFP: Spent Fuel

Studies a t the'Barnwel1 Nuclear Plant," Allied General Nuclear Services,

Barnwell, SC, AGNS-35900-1.1-183.

Baxter, R.G. , 1983, !'Description of Defense Waste Processing Fac i l i t y

Reference Waste Form and Canister," E.I. du Pont de Nemours & Co.,

Savannah River Plant, Aiken, SC., DP-1606 Rev. 1 , August 1983.

Burns, Patrick J., 1982, "TACOZD, A Finite Element Heat Transfer Code,"

Lawrence Livermore National Laboratory, Livermore, CA., UCID-17980, Rev. 2.

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Dudley, W . W . , J r . , and Erdal, B.R., 1982, "Si te Characterization f o r

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McCright , R . D., Weiss, H. , Juhas, M.C.., and Logan, R.W., 1983, IISelection

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Russel 1 , E.W., McCrignt, R.D. and O'Neal , W.C. , 1983, IIContainment Barrier

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Schornhorst, J.R. e t a l , 1983, "Conceptual Waste Package Designs f o r

Disposal of Nuclear Waste in Tuff," Westinghouse AESD, Pittsburgh, PA,

ONWI-439, Bat te l le , Office of Nuclear Waste Isolat ion, Columbus, OH.

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I